Pub Date : 2024-09-27DOI: 10.1016/j.pnucene.2024.105463
The encapsulation of Radioactive Reactive Metallic Waste (RRMW) in ordinary Portland cement poses significant challenges due to its incompatibility with the alkaline environment of the matrix. To address this issue, magnesium phosphate cements (MPC) emerge as potential solutions for the safe and effective immobilisation of RRMWs. The radiation stability and durability of an optimised formulation have been examined for samples irradiated up to 1000 kGy, in particular concerning the leaching behaviour of the three main constituents of the cement hydration products, and on four artificially added elements used to simulate radionuclides commonly found in radioactive waste (caesium, strontium, europium, and cobalt). The mortars exhibited excellent leaching behaviour and a high mechanical resistance, even after irradiation, freeze-thaw cycles, and water immersion. No significant radiation-induced effects were observed in the mineralogical and microstructural properties of the mortars, thus supporting their stability at the examined doses. Having verified the compliance with the main Italian waste acceptance criteria, the results of this research represent an encouraging step for the future implementation of MPCs for RRMWs conditioning.
{"title":"Radiation stability and durability of magnesium phosphate cement for radioactive reactive metals encapsulation","authors":"","doi":"10.1016/j.pnucene.2024.105463","DOIUrl":"10.1016/j.pnucene.2024.105463","url":null,"abstract":"<div><div>The encapsulation of Radioactive Reactive Metallic Waste (RRMW) in ordinary Portland cement poses significant challenges due to its incompatibility with the alkaline environment of the matrix. To address this issue, magnesium phosphate cements (MPC) emerge as potential solutions for the safe and effective immobilisation of RRMWs. The radiation stability and durability of an optimised formulation have been examined for samples irradiated up to 1000 kGy, in particular concerning the leaching behaviour of the three main constituents of the cement hydration products, and on four artificially added elements used to simulate radionuclides commonly found in radioactive waste (caesium, strontium, europium, and cobalt). The mortars exhibited excellent leaching behaviour and a high mechanical resistance, even after irradiation, freeze-thaw cycles, and water immersion. No significant radiation-induced effects were observed in the mineralogical and microstructural properties of the mortars, thus supporting their stability at the examined doses. Having verified the compliance with the main Italian waste acceptance criteria, the results of this research represent an encouraging step for the future implementation of MPCs for RRMWs conditioning.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":3.3,"publicationDate":"2024-09-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142326765","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-09-26DOI: 10.1016/j.pnucene.2024.105450
This study presents a computational capability for fission product retention and release in two-phase, multi-species systems representing Molten Salt Reactors (MSR) with coupled thermal-hydraulics and fuel coolant chemical behaviours. This is demonstrated through four simulated cases centred on the proposed Molten Salt Fast Reactor (MSFR). This is achieved by two-way coupling the Computational Fluid Dynamics (CFD) code OpenFOAM and the Computational Thermodynamics (CT) code Thermochimica, using the Joint Research Centre Molten Salt Database (JRCMSD). Local chemical equilibrium is assumed, implying that chemical kinetics are predominantly governed by mass transport. Four simulations address normal operating conditions, exploring: (i) dilution of fission products injected within the molten salt coolant, (ii) molten salt coolant evaporation rate, (iii) release of radioactive gaseous species, (iv) shifts in the UF4/UF3 ratio, and (v) comparison of vapour pressures of gaseous species. The influence of temperature-dependent viscosity on retaining fission products, compared to consistent values, is also discussed. The feasibility of integrating CFD with Thermochimica showed promising results, broadening insights into multiphysics systems and setting the stage for its application in more intricate scenarios.
{"title":"Coupled computational fluid dynamics and computational thermodynamics simulations for fission product retention and release: A molten salt fast reactor application","authors":"","doi":"10.1016/j.pnucene.2024.105450","DOIUrl":"10.1016/j.pnucene.2024.105450","url":null,"abstract":"<div><div>This study presents a computational capability for fission product retention and release in two-phase, multi-species systems representing Molten Salt Reactors (MSR) with coupled thermal-hydraulics and fuel coolant chemical behaviours. This is demonstrated through four simulated cases centred on the proposed Molten Salt Fast Reactor (MSFR). This is achieved by two-way coupling the Computational Fluid Dynamics (CFD) code <span>OpenFOAM</span> and the Computational Thermodynamics (CT) code <span>Thermochimica</span>, using the Joint Research Centre Molten Salt Database (JRCMSD). Local chemical equilibrium is assumed, implying that chemical kinetics are predominantly governed by mass transport. Four simulations address normal operating conditions, exploring: (i) dilution of fission products injected within the molten salt coolant, (ii) molten salt coolant evaporation rate, (iii) release of radioactive gaseous species, (iv) shifts in the UF<sub>4</sub>/UF<sub>3</sub> ratio, and (v) comparison of vapour pressures of gaseous species. The influence of temperature-dependent viscosity on retaining fission products, compared to consistent values, is also discussed. The feasibility of integrating CFD with Thermochimica showed promising results, broadening insights into multiphysics systems and setting the stage for its application in more intricate scenarios.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":3.3,"publicationDate":"2024-09-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142323523","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-09-26DOI: 10.1016/j.pnucene.2024.105462
Helium-xenon cooled reactor direct Brayton cycle system has excellent application prospects in small nuclear power plants due to its light weight, high compactness and simple structure. The safe startup of reactor system is very important, so the system startup scheme is designed. In addition, reactor system may be affected by external acceleration during operation. Whether reactor system can operate normally and safely after being affected by acceleration and its influence characteristics are not clear. In order to verify the feasibility of the system startup scheme and explore the acceleration influence characteristics, the system analysis program was developed to simulate the system startup transient operating conditions and the transient operating conditions of reactor system under the influence of acceleration. In the research of acceleration influence characteristics, acceleration has different periods, amplitudes, types and directions, and the load state is divided into load change with system output power and constant load. The results show that the designed system startup scheme can start reactor system safely. Reactor system can ignore the influence of acceleration when it has load change with system output power ability. If the load remains constant, reactor system cannot ensure continuous and stable operation after being affected by acceleration, which has the risk of Loss of Flow Accident (LOFA). The relevant research results provide theoretical reference and support for the design of the reactor system startup scheme and the research on the influence characteristics of acceleration on reactor system.
{"title":"Preliminary analysis of startup and acceleration transient characteristics of helium-xenon cooled reactor direct brayton cycle system","authors":"","doi":"10.1016/j.pnucene.2024.105462","DOIUrl":"10.1016/j.pnucene.2024.105462","url":null,"abstract":"<div><div>Helium-xenon cooled reactor direct Brayton cycle system has excellent application prospects in small nuclear power plants due to its light weight, high compactness and simple structure. The safe startup of reactor system is very important, so the system startup scheme is designed. In addition, reactor system may be affected by external acceleration during operation. Whether reactor system can operate normally and safely after being affected by acceleration and its influence characteristics are not clear. In order to verify the feasibility of the system startup scheme and explore the acceleration influence characteristics, the system analysis program was developed to simulate the system startup transient operating conditions and the transient operating conditions of reactor system under the influence of acceleration. In the research of acceleration influence characteristics, acceleration has different periods, amplitudes, types and directions, and the load state is divided into load change with system output power and constant load. The results show that the designed system startup scheme can start reactor system safely. Reactor system can ignore the influence of acceleration when it has load change with system output power ability. If the load remains constant, reactor system cannot ensure continuous and stable operation after being affected by acceleration, which has the risk of Loss of Flow Accident (LOFA). The relevant research results provide theoretical reference and support for the design of the reactor system startup scheme and the research on the influence characteristics of acceleration on reactor system.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":3.3,"publicationDate":"2024-09-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142323522","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-09-26DOI: 10.1016/j.pnucene.2024.105464
A scoping study was performed for chlorinating dross formed during uranium casting operations. The purpose is to minimize the losses of uranium to dross wastes. The dross is primarily a mixture of uranium metal and uranium oxide with a minor fraction of crucible and crucible coating materials. The reaction chemistries were performed in a carrier salt of LiCl-KCl eutectic at 500 °C. The addition of FeCl2 chlorinated the uranium metal to UCl3, by reducing the FeCl2 to iron metal. After the uranium metal was chlorinated, zirconium metal was added to the salt. The residual FeCl2 chlorinated the zirconium metal to ZrCl4, by reducing the FeCl2 to iron metal. In turn, the ZrCl4 chlorinated the uranium oxide to UCl3, by converting the ZrCl4 to zirconium oxide. The effectiveness of the chlorination reactions was qualitatively verified by cyclic voltammograms that indicated the presence or absence of FeCl2 and UCl3 in the salt.
{"title":"Chlorination of HALEU regulus casting dross","authors":"","doi":"10.1016/j.pnucene.2024.105464","DOIUrl":"10.1016/j.pnucene.2024.105464","url":null,"abstract":"<div><div>A scoping study was performed for chlorinating dross formed during uranium casting operations. The purpose is to minimize the losses of uranium to dross wastes. The dross is primarily a mixture of uranium metal and uranium oxide with a minor fraction of crucible and crucible coating materials. The reaction chemistries were performed in a carrier salt of LiCl-KCl eutectic at 500 °C. The addition of FeCl<sub>2</sub> chlorinated the uranium metal to UCl<sub>3</sub>, by reducing the FeCl<sub>2</sub> to iron metal. After the uranium metal was chlorinated, zirconium metal was added to the salt. The residual FeCl<sub>2</sub> chlorinated the zirconium metal to ZrCl<sub>4</sub>, by reducing the FeCl<sub>2</sub> to iron metal. In turn, the ZrCl<sub>4</sub> chlorinated the uranium oxide to UCl<sub>3</sub>, by converting the ZrCl<sub>4</sub> to zirconium oxide. The effectiveness of the chlorination reactions was qualitatively verified by cyclic voltammograms that indicated the presence or absence of FeCl<sub>2</sub> and UCl<sub>3</sub> in the salt.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":3.3,"publicationDate":"2024-09-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142323521","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-09-26DOI: 10.1016/j.pnucene.2024.105459
Advanced pressurized water reactor was simulated using system analysis program. The depressurization strategy for the automatic depressurization system (ADS) under typical ADS-triggering accidents (2-inch/10-inch cold leg break, inadvertent opening of the ADS, DVI pipeline break) was investigated. The results show that after these accidents, ADS can ensure the operation of the three types of safe injection tanks, with high redundancy and inherent safety. The multi-stage design of ADS aids the sparger in maintaining a stable critical jet state. In 10-inch cold leg break accident, the sparger cannot reached the critical jet state under all of pipeline fault conditions. From a system point of view, adjusting the pipeline design of ADS is necessary to avoid long-term condensation oscillation of the sparger in some accidents. From a design perspective of the local equipment, the nozzle structure of the sparger should attenuate the dynamic load caused by the condensation oscillation and back-attack phenomenon.
{"title":"Depressurization strategy of China's advanced pressurized water reactor under typical accidents","authors":"","doi":"10.1016/j.pnucene.2024.105459","DOIUrl":"10.1016/j.pnucene.2024.105459","url":null,"abstract":"<div><div>Advanced pressurized water reactor was simulated using system analysis program. The depressurization strategy for the automatic depressurization system (ADS) under typical ADS-triggering accidents (2-inch/10-inch cold leg break, inadvertent opening of the ADS, DVI pipeline break) was investigated. The results show that after these accidents, ADS can ensure the operation of the three types of safe injection tanks, with high redundancy and inherent safety. The multi-stage design of ADS aids the sparger in maintaining a stable critical jet state. In 10-inch cold leg break accident, the sparger cannot reached the critical jet state under all of pipeline fault conditions. From a system point of view, adjusting the pipeline design of ADS is necessary to avoid long-term condensation oscillation of the sparger in some accidents. From a design perspective of the local equipment, the nozzle structure of the sparger should attenuate the dynamic load caused by the condensation oscillation and back-attack phenomenon.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":3.3,"publicationDate":"2024-09-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142323524","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-09-26DOI: 10.1016/j.pnucene.2024.105460
The qualification of nuclear fuel is the process by which a fuel is certified for use within a reactor design. This process involves massive technical and procedural coordination across multiple private and public institutions with long project completion timelines. The current frameworks published in literature do not address the roles of private reactor developers in the qualification process, nor the regulatory requirements that form the background for fuel qualification. These roles and requirements are not static and have evolved throughout the development of the commercial nuclear industry. UO2 fuel, the only fuel considered fully qualified today, achieved its status in large part due to its early selection and rapid deployment under a still developing regulatory landscape. Today, fuel qualification efforts must operate within modern regulatory realities, which will require extensive testing and model development. Building on the existing fuel qualification frameworks, we define the optimal roles of the parties involved with fuel qualification. We also suggest viewing experimental efforts through a data-centric perspective in which parallel improvements can be made in collection, processing, and storage to facilitate quicker qualification timelines. These efforts then promote advanced modeling efforts, such as mechanistic modeling and digital twin development.
{"title":"Nuclear fuel qualification: History, current state, and future","authors":"","doi":"10.1016/j.pnucene.2024.105460","DOIUrl":"10.1016/j.pnucene.2024.105460","url":null,"abstract":"<div><div>The qualification of nuclear fuel is the process by which a fuel is certified for use within a reactor design. This process involves massive technical and procedural coordination across multiple private and public institutions with long project completion timelines. The current frameworks published in literature do not address the roles of private reactor developers in the qualification process, nor the regulatory requirements that form the background for fuel qualification. These roles and requirements are not static and have evolved throughout the development of the commercial nuclear industry. UO<sub>2</sub> fuel, the only fuel considered fully qualified today, achieved its status in large part due to its early selection and rapid deployment under a still developing regulatory landscape. Today, fuel qualification efforts must operate within modern regulatory realities, which will require extensive testing and model development. Building on the existing fuel qualification frameworks, we define the optimal roles of the parties involved with fuel qualification. We also suggest viewing experimental efforts through a data-centric perspective in which parallel improvements can be made in collection, processing, and storage to facilitate quicker qualification timelines. These efforts then promote advanced modeling efforts, such as mechanistic modeling and digital twin development.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":3.3,"publicationDate":"2024-09-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142323520","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-09-25DOI: 10.1016/j.pnucene.2024.105448
There have been considerable advances of shutdown dose rate (SDDR) calculation methods for fusion related problems, however applications of these methods in the fission reactor field have hitherto been sparse. The present study attempts to bridge this gap by investigating the applicability of SDDR calculation methods for fission reactor problems. Specifically, we aim to assess and validate whether recent advances in SDDR methods can be successfully applied in fission research reactors. To this end, we estimate the shutdown dose rate distribution at the Open Pool Australian Light water reactor (OPAL) using the rigorous two step (R2S) computational method, and we compare the calculated results with the experimental data. This method utilizes a 3D reactor model implemented in the Monte Carlo N-Particle (MCNP) transport code, the AutomeD VAriaNce reduction Generator (ADVANTG) code for geometry discretization and variance reduction calculations, and the Oak Ridge Isotope GENeration (ORIGEN) inventory code for activation calculations. To ensure robustness, we employ two variance reduction techniques, Forward Weighted Consistent Adjoint Driven Importance Sampling (FW-CADIS) and Multi-Step Consistent Adjoint Driven Importance Sampling (MS-CADIS). To the best of our knowledge, this is the first MS-CADIS method implementation for fission reactor problems. The SDDR is estimated at ten locations within the experimental hall, all situated more than 4 m away from the reactor core.
The paper shows that, the experimental observations are within the lower and upper bounds of the simulation results for 4 out of 10 locations, while the remaining observations are within a factor of 7, with one significant outlier. The calculated average dose rate is within 5% of the nominal values of the experimental observations for 3 locations. The computational results are within statistical uncertainty by using two different variance reduction techniques, with significant computational advantage of MS-CADIS over FW-CADIS for SDDR calculations. The results indicate that the combination of SDDR distribution maps, estimated dose rate energy dependance, and activation information are powerful tools in identifying the radioisotopes and reactor components dominating the SDDR. These results can contribute to better radiation safety practices in contaminated areas, by enabling the minimal dose path planning or by improving radiation shielding.
{"title":"Shutdown dose rate calculation for fission reactors: An application of the MS-CADIS method to OPAL","authors":"","doi":"10.1016/j.pnucene.2024.105448","DOIUrl":"10.1016/j.pnucene.2024.105448","url":null,"abstract":"<div><div>There have been considerable advances of shutdown dose rate (SDDR) calculation methods for fusion related problems, however applications of these methods in the fission reactor field have hitherto been sparse. The present study attempts to bridge this gap by investigating the applicability of SDDR calculation methods for fission reactor problems. Specifically, we aim to assess and validate whether recent advances in SDDR methods can be successfully applied in fission research reactors. To this end, we estimate the shutdown dose rate distribution at the Open Pool Australian Light water reactor (OPAL) using the rigorous two step (R2S) computational method, and we compare the calculated results with the experimental data. This method utilizes a 3D reactor model implemented in the Monte Carlo N-Particle (MCNP) transport code, the AutomeD VAriaNce reduction Generator (ADVANTG) code for geometry discretization and variance reduction calculations, and the Oak Ridge Isotope GENeration (ORIGEN) inventory code for activation calculations. To ensure robustness, we employ two variance reduction techniques, Forward Weighted Consistent Adjoint Driven Importance Sampling (FW-CADIS) and Multi-Step Consistent Adjoint Driven Importance Sampling (MS-CADIS). To the best of our knowledge, this is the first MS-CADIS method implementation for fission reactor problems. The SDDR is estimated at ten locations within the experimental hall, all situated more than 4 m away from the reactor core.</div><div>The paper shows that, the experimental observations are within the lower and upper bounds of the simulation results for 4 out of 10 locations, while the remaining observations are within a factor of 7, with one significant outlier. The calculated average dose rate is within 5% of the nominal values of the experimental observations for 3 locations. The computational results are within statistical uncertainty by using two different variance reduction techniques, with significant computational advantage of MS-CADIS over FW-CADIS for SDDR calculations. The results indicate that the combination of SDDR distribution maps, estimated dose rate energy dependance, and activation information are powerful tools in identifying the radioisotopes and reactor components dominating the SDDR. These results can contribute to better radiation safety practices in contaminated areas, by enabling the minimal dose path planning or by improving radiation shielding.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":3.3,"publicationDate":"2024-09-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142319936","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-09-24DOI: 10.1016/j.pnucene.2024.105447
In 2018, the NECP Laboratory at Xi'an Jiaotong University (XJTU) published the pin-by-pin two-step calculation code system NECP-Bamboo2.0 for the physics analysis of Pressurized Water Reactors (PWRs). Previously, NECP-Bamboo2.0 was primarily verified for its neutronics analysis capabilities. The code system has recently been enhanced for practical reactor physics analysis of commercial PWR cores. In this paper, several modifications of NECP-Bamboo2.0 regarding its improvement in functionality are introduced. Then, the pin-by-pin calculation capability of the code system was verified using the 2D and 3D VERA benchmark problems. Comparisons with reference solutions provided by high-fidelity codes demonstrated the computational accuracy of NECP-Bamboo2.0. Additionally, NECP-Bamboo2.0 was applied to the multi-cycle physics analysis of two commercial PWR cores, CNP300 and CNP1000. Comparisons with the measurements regarding the startup physics tests and the depletion processes illustrate the reliability of NECP-Bamboo2.0 in practical pin-by-pin reactor physics analysis and highlight the improvements of the pin-by-pin method over the assembly-homogenized coarse-mesh diffusion method.
2018年，西安交通大学NECP实验室发布了用于压水堆物理分析的逐针两步计算代码系统NECP-Bamboo2.0。此前，NECP-Bamboo2.0 主要验证了其中子分析能力。最近，该代码系统针对商用压水堆堆芯的实际反应堆物理分析进行了增强。本文介绍了NECP-Bamboo2.0在功能改进方面的几处修改。然后，使用二维和三维 VERA 基准问题验证了代码系统的逐针计算能力。通过与高保真代码提供的参考解进行比较，证明了NECP-Bamboo2.0的计算精度。此外，NECP-Bamboo2.0还应用于CNP300和CNP1000两个商用压水堆堆芯的多周期物理分析。与有关启动物理试验和耗竭过程的测量结果的比较说明了NECP-Bamboo2.0在实际逐针反应堆物理分析中的可靠性，并突出了逐针方法与装配均质化粗网格扩散方法相比的改进之处。
{"title":"Verifications and applications of NECP-Bamboo2.0 for PWR whole-core multi-cycle pin-by-pin simulation","authors":"","doi":"10.1016/j.pnucene.2024.105447","DOIUrl":"10.1016/j.pnucene.2024.105447","url":null,"abstract":"<div><div>In 2018, the NECP Laboratory at Xi'an Jiaotong University (XJTU) published the pin-by-pin two-step calculation code system NECP-Bamboo2.0 for the physics analysis of Pressurized Water Reactors (PWRs). Previously, NECP-Bamboo2.0 was primarily verified for its neutronics analysis capabilities. The code system has recently been enhanced for practical reactor physics analysis of commercial PWR cores. In this paper, several modifications of NECP-Bamboo2.0 regarding its improvement in functionality are introduced. Then, the pin-by-pin calculation capability of the code system was verified using the 2D and 3D VERA benchmark problems. Comparisons with reference solutions provided by high-fidelity codes demonstrated the computational accuracy of NECP-Bamboo2.0. Additionally, NECP-Bamboo2.0 was applied to the multi-cycle physics analysis of two commercial PWR cores, CNP300 and CNP1000. Comparisons with the measurements regarding the startup physics tests and the depletion processes illustrate the reliability of NECP-Bamboo2.0 in practical pin-by-pin reactor physics analysis and highlight the improvements of the pin-by-pin method over the assembly-homogenized coarse-mesh diffusion method.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":3.3,"publicationDate":"2024-09-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142314724","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-09-24DOI: 10.1016/j.pnucene.2024.105451
This study presents the optical and radiation shielding characteristics of a newly developed borate glass, doped with equal quantities of ZnO and BaO, and modified with calcium and sodium. The glass samples were prepared using the melt quenching technique, with the composition formula (80-x-y)B2O3-10CaO-10Na2O-xZnO-yBaO, where x and y were varied at 5, 10, 15, and 20 mol%. Comprehensive analyses of the optical and gamma shielding properties were carried out using UV–visible spectrophotometry and Phy-X software, respectively. The UV–visible spectroscopic data allowed for the calculation of various parameters including the direct and indirect optical energy band gaps, refractive index, dielectric constants, and polarizability. It was observed that the direct optical energy band gap decreased from 3.91 to 3.10 eV, while the indirect band gap fell from 3.41 to 2.91 eV with increasing ZnO and BaO content. Conversely, the refractive index rose from 2.29 to 2.42 with higher concentrations of ZnO and BaO. The linear attenuation coefficients (LACs) across the range of 0.015–15 MeV exhibited an inverse relationship. Among the glass samples, B40Zn20Ba20 exhibited the lowest tenth value layer (TVL) and the highest effective atomic number (Zeff), indicating its superior performance as a radiation shielding material. Overall, the produced glass samples demonstrate commendable properties for both optical applications and radiation shielding.
本研究介绍了一种新开发的掺杂了等量氧化锌和氧化钡并用钙和钠改性的硼酸盐玻璃的光学和辐射屏蔽特性。玻璃样品采用熔体淬火技术制备，其组成式为 (80-x-y)B2O3-10CaO-10Na2O-xZnO-yBaO ，其中 x 和 y 分别为 5、10、15 和 20 摩尔%。紫外可见分光光度法和 Phy-X 软件分别对光学和伽马屏蔽特性进行了全面分析。紫外可见光谱数据可用于计算各种参数，包括直接和间接光学能带隙、折射率、介电常数和极化性。结果表明，随着氧化锌和氧化钡含量的增加，直接光能带隙从 3.91 eV 下降到 3.10 eV，而间接能带隙则从 3.41 eV 下降到 2.91 eV。相反，随着 ZnO 和 BaO 含量的增加，折射率从 2.29 上升到 2.42。在 0.015-15 MeV 的范围内，线性衰减系数（LAC）呈反比关系。在玻璃样品中，B40Zn20Ba20 的十值层（TVL）最低，有效原子序数（Zeff）最高，表明其作为辐射屏蔽材料的性能优越。总体而言，所生产的玻璃样品在光学应用和辐射屏蔽方面都表现出了值得称道的性能。
{"title":"Optical and gamma-ray shielding properties of calcium sodium borate glasses with varied equal concentrations of ZnO and BaO","authors":"","doi":"10.1016/j.pnucene.2024.105451","DOIUrl":"10.1016/j.pnucene.2024.105451","url":null,"abstract":"<div><div>This study presents the optical and radiation shielding characteristics of a newly developed borate glass, doped with equal quantities of ZnO and BaO, and modified with calcium and sodium. The glass samples were prepared using the melt quenching technique, with the composition formula (80-x-y)B<sub>2</sub>O<sub>3</sub>-10CaO-10Na<sub>2</sub>O-xZnO-yBaO, where x and y were varied at 5, 10, 15, and 20 mol%. Comprehensive analyses of the optical and gamma shielding properties were carried out using UV–visible spectrophotometry and Phy-X software, respectively. The UV–visible spectroscopic data allowed for the calculation of various parameters including the direct and indirect optical energy band gaps, refractive index, dielectric constants, and polarizability. It was observed that the direct optical energy band gap decreased from 3.91 to 3.10 eV, while the indirect band gap fell from 3.41 to 2.91 eV with increasing ZnO and BaO content. Conversely, the refractive index rose from 2.29 to 2.42 with higher concentrations of ZnO and BaO. The linear attenuation coefficients (LACs) across the range of 0.015–15 MeV exhibited an inverse relationship. Among the glass samples, B40Zn20Ba20 exhibited the lowest tenth value layer (TVL) and the highest effective atomic number (Zeff), indicating its superior performance as a radiation shielding material. Overall, the produced glass samples demonstrate commendable properties for both optical applications and radiation shielding.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":3.3,"publicationDate":"2024-09-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://www.sciencedirect.com/science/article/pii/S0149197024004013/pdfft?md5=cee1d69d9886f1635b396a7acef525ec&pid=1-s2.0-S0149197024004013-main.pdf","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142312488","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-09-24DOI: 10.1016/j.pnucene.2024.105456
One of the primary challenges in negotiating treaties related to the control of Nuclear Materials (NM) is the verification process, particularly when dealing with sensitive information. Normally, in this process, the inspector utilizes a suitable measuring system to verify the declared information. The difficulty may arise whenever sensitive information related to the NM should not be released to an inspector. Simultaneously, the NM owner (operator) should remain unaware of the measuring system employed by the inspector to prevent any probable manipulation. To address this issue, a neutral third party is assumed to act as an intermediary, who matches the data declared by the operator with the results obtained by the inspector, without exchanging any information between them. In this work, a Non-Intrusive System (NIS) is proposed and tested to play this role. The system receives data and information from both the operator and the inspector in the form of CAD or Monte Carlo (MC) input files, in addition to the results of measurements performed by the inspector. Then the system performs calculations and combines the results of these calculations with the inspector's measurement result to estimate the mass of the NM. The only information automatically conveyed to the inspector is the final conclusion regarding whether the declared and the estimated masses of NM are matched or not. The proposed NIS concept is tested using samples of NMs, and the results are presented.
{"title":"A Proposed Non-Intrusive System for the Verification of Sensitive Nuclear Material","authors":"","doi":"10.1016/j.pnucene.2024.105456","DOIUrl":"10.1016/j.pnucene.2024.105456","url":null,"abstract":"<div><div>One of the primary challenges in negotiating treaties related to the control of Nuclear Materials (NM) is the verification process, particularly when dealing with sensitive information. Normally, in this process, the inspector utilizes a suitable measuring system to verify the declared information. The difficulty may arise whenever sensitive information related to the NM should not be released to an inspector. Simultaneously, the NM owner (operator) should remain unaware of the measuring system employed by the inspector to prevent any probable manipulation. To address this issue, a neutral third party is assumed to act as an intermediary, who matches the data declared by the operator with the results obtained by the inspector, without exchanging any information between them. In this work, a Non-Intrusive System (NIS) is proposed and tested to play this role. The system receives data and information from both the operator and the inspector in the form of CAD or Monte Carlo (MC) input files, in addition to the results of measurements performed by the inspector. Then the system performs calculations and combines the results of these calculations with the inspector's measurement result to estimate the mass of the NM. The only information automatically conveyed to the inspector is the final conclusion regarding whether the declared and the estimated masses of NM are matched or not. The proposed NIS concept is tested using samples of NMs, and the results are presented.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":3.3,"publicationDate":"2024-09-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142314722","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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